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A significant role for a future nuclear carbon-free energy production is attributed to fast reactors, mostly employing a liquid metal as a coolant. This paper summarizes the efforts that have been undertaken in collaborative projects sponsored by the European Commission in the past 20 years in the fields of liquid-metal heat transfer modeling, fuel assembly and core thermal hydraulics, pool and system thermal hydraulics, and establishment of best practice guidelines and verification, validation, and uncertainty quantification (UQ). The achievements in these fields will be presented along with the prospects on topics which will be studied collaboratively in Europe in the years to come. These prospects include further development of heat transfer models for applied computational fluid dynamics (CFD), further analysis of the consequences of fuel assembly blockages on coolant flow and temperature, analysis of the thermal hydraulic effects in deformed fuel assemblies, extended validation of three-dimensional pool thermal hydraulic CFD models, and further development and validation of multi-scale system thermal hydraulic methods.

Heavy-liquid metals (HLMs), such as lead and lead–bismuth eutectic (LBE), are proposed as primary coolants in accelerator driven systems and next-generation fast reactors. In Europe, the reference systems using HLMs are MYRRHA (LBE) and ALFRED (lead). Extensive R&D programs have been established for supporting their detailed design and safety assessment, including thermal–hydraulic experiments at representative operating conditions in an HLM environment. These experiments aim both at a design verification and at the validation of numerical models, which allow an extrapolation of the results. Advanced instrumentation, capable of sustaining high temperatures and corrosion, is necessary for accurate measurements, often in compact geometries. This article presents an overview of recent experiences and ongoing activities on pool-type and loop-type HLM experiments. Pool tests include the measurement of forced- and natural-circulation flow patterns in several scenarios representative of nominal and decay heat removal conditions. Loop tests are focused on the evaluation of specific components, like mockups of the fuel assembly, control rod and heat exchangers. They involve the measurement of global variables, such as flow rate and pressure difference, and local quantities like temperature, velocity and vibrations. In addition to traditional techniques, other instrumentation based on optical fibers, ultrasonic and electromagnetic methods are discussed.

The thermal-hydraulic challenges of a nuclear reactor are numerous and mastering them is crucial for the design and safety of new reactors. Numerical simulation through computational fluid dynamics...

This work on fuel dispersion aims at quantifying the design and safety of the MYRRHA nuclear reactor. A number of accidents leading to the release of a secondary phase into the primary coolant loop are investigated. Among these scenarios, an incident leading to the failure of one or more of the fuel pins is simulated while the reactor is operating in nominal conditions, but also in natural convection regime either during accident transients such as loss-of-flow or during the normal shut-down of the reactor. Two single-phase CFD models of the MYRRHA reactor are constructed in ANSYS Fluent to represent the reactor in nominal and natural convection conditions. An Euler–Lagrange approach with one-way coupling is used for the flow and particle tracking. Firstly, a steady state RANS solution is obtained for each of the three conditions. Secondly, the particles are released downstream from the core outlet and particle distributions are provided over the coolant circuit. Their size and density are defined such that test cases represent potential extremes that may occur. Analysis of the results highlights different particle behaviors, depending essentially on gravity forces and kinematic effects. Statistical distributions highlight potential accumulation regions that may form at the free-surfaces, on top of the upper diaphragm plate or at the bottom of the vessel. These results help to localize regions of fuel accumulation in order to provide insight for development of strategies for accident mitigation.

The European Scaled Pool Experiment (E-SCAPE) is a 1/6th-scale thermal hydraulic model of the MYRRHA reactor and has been developed within the European framework programs THINS and MYRTE. MYRRHA is a multi-purpose hybrid pool-type research reactor cooled with lead-bismuth eutectic (LBE) that is under design at the Belgian nuclear research center SCK•CEN. The experiments from the E-SCAPE facility are essential for the design and licensing of MYRRHA. On the one hand, the experimental data from E-SCAPE directly provide information to the designers on the flow patterns and heat transport in the LBE in the upper and lower plena of the reactor. On the other hand, the experimental data from E-SCAPE enable qualification of thermal hydraulic computer codes used for analyzing the flow and heat transport in the MYRRHA reactor. The accuracy of codes rely on the underlying modelling approaches, whose validity can only be assessed by comparing them with relevant experimental data. Given the rare conditions present in heavy liquid cooled nuclear reactors, the possibility to compare complete pool simulations with experimental data is a unique opportunity. This paper presents the modelling approaches and first results of the pre-test analyses performed for E-SCAPE by NRG, SCK•CEN and VKI using Computational Fluid Dynamics (CFD). The CFD analyses of NRG are performed with the STAR-CCM + code, at SCK•CEN with the ANSYS CFX code and at VKI OpenFOAM is used. A first comparison of the results obtained with three different CFD codes for flow and heat transport in the E-SCAPE pool is presented and, where possible, validated against experimental results.

MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) is a flexible fast-spectrum research reactor under design at SCK·CEN. MYRRHA is a pool-type reactor with lead bismuth eutectic (LBE) as primary coolant. The proper understanding of the thermal hydraulic phenomena occurring in the reactor pool is an important issue in the design and licensing of the MYRRHA system and liquid-metal cooled reactors by extension. Model experiments are necessary for understanding the physics, for validating experimental tools and to qualify the design for the licensing. The E-SCAPE (European SCAled Pool Experiment) facility at SCK·CEN is a thermal hydraulic 1/6-scale model of the MYRRHA reactor, with an electrical core simulator, cooled by LBE. It provides experimental feedback to the designers on the forced and natural circulation flow patterns. Moreover, it enables to validate the computational methods for their use with LBE. The paper will elaborate on the design of the E-SCAPE facility and its main parameters. Also the experimental matrix and the pre-test analysis using computational fluid dynamics (CFD) and system thermal hydraulics codes will be described.

Vibration measurements of the fuel assembly of a nuclear reactor are a very useful tool to determine the health and lifetime of the reactor core. The importance of these measurements is exacerbated in the new generation of heavy liquid metal reactors, where the fuel assembly is exposed to a corrosive molten metal coolant at 300 $^\circ$ C and where the space between the individual fuel pins is limited to a few millimeters. In this paper, we consider fibre Bragg gratings as potential candidates for carrying out fuel pin vibration measurements in such an environment. We describe a dedicated method to integrate fibre Bragg gratings in a fuel pin, and we subject this pin to conditions close to those encountered in a real heavy liquid metal reactor. More specifically, we report on the performance of draw tower gratings used as a vibration sensor when the fuel pins are immersed in heavy liquid metal at 300 $^\circ$ C for up to 700 h. The performance evaluation is based on monitoring the signal-to-noise ratio of the grating's spectral response as a function of time. We show that accurate detection of the Bragg peak becomes very challenging after 400 h of exposure. Additionally, we succeed to extend the useful lifetime with a factor of two by using an appropriate integration of the fiber in the fuel pin and by using an alternate peak detection algorithm.

International audience; In this article, the fluid forces and the dynamics of a flexible clamped-clamped cylinder in turbulent axial flow are computed numerically. In the presented numerical model, there is no need to tune parameters for each specific case or to obtain coefficients from experiments. The results are compared with the dynamics measured in experiments available in literature. The specific case studied here consists of a silicone cylinder mounted in axial water flow. Computationally it is found that the cylinder loses stability first by buckling. The threshold for buckling is in quantitative agreement with experimental results and weakly-nonlinear theory. At higher flow speed a fluttering motion is predicted, in agreement with experimental results. It is also shown that even a small misalignment between the flow and the structure can have a significant impact on the dynamical behavior. To provide insight in the results of these fluid-structure interaction simulations, forces are computed on rigid inclined and curved cylinders, showing the existence of two different flow regimes. Furthermore it is shown that the inlet turbulence state has a non-negligible effect on these forces and thus on the dynamics of the cylinder.

Excessive vibration of nuclear reactor components, such as the heat exchanger or the fuel assembly should be avoided as these can compromise the lifetime of these components and potentially lead to safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants. However, identifying adequate sensors or techniques that can be successfully applied to record the vibrations of the components in a flow of liquid metal at elevated temperatures is very challenging. In this paper, we demonstrate the precise measurements of the vibrations of a very representative mock-up of a fuel assembly in a lead-bismuth eutectic cooled installation using quasi-distributed fibre Bragg grating (FBG) based sensors. The unique properties of these sensors, in combination with a dedicated integration and mounting approach, allows for accounting of the severe geometrical constraints and allows characterizing the vibration of the fuel assembly elements under nominal operation conditions. To that aim, we instrumented a single fuel pin within the fuel assembly with 84 FBGs, and conducted spectral measurements with an acquisition rate of up to 5000 measurements per second, enabling the monitoring of local strains of a few mu epsilon. These measurements provide the information required to assess vibration-related safety hazards.

Liquid metal fast reactors (LMFRs) are foreseen to play an important role in the future of nuclear energy, thanks to their increased fuel utilization and safety features profiting from the optimal heat transfer performance of the metallic coolants. Accurate thermal-hydraulic analysis of their fuel assemblies, typically employed with wire-wraps as spacers, is recognized as a crucial scientific and engineering contribution to support the deployment of such technology. This challenges the modeling and simulation community. To this aspect, various reference databases (both experimental and numerical) for different wire-wrapped fuel assembly configurations have been created recently and are being used for validation of engineering simulation approaches based on Reynolds Averaged Navier Stokes (RANS) modelling. These databases include: • 7-pin rod bundle: A detailed experiment with Particle Image Velocimetry (PIV) is performed. In order to allow accurate measurements of the flow topology, a matched-index-of-refraction technique was used employing water as working fluid. • 19-pin bundle: A series of experiments is performed covering a wide range of Reynolds and Peclet numbers as well as thermal powers. The experiments use liquid lead-bismuth eutectic as working fluid. The measurements include pressure drop and local temperatures. • 61-pin rod bundle: This large eddy simulation including conjugate heat transfer from the pin cladding to the coolant allows to bridge the gap from small bundles (less than 37 pins) to large bundles (more than 37 pins). In literature, a fundamental different behavior has been observed for small bundles compared to large bundles. • 127-pin bundle: Isothermal experiments using lead-bismuth eutectic characterizing pressure drop are performed on a full scale fuel assembly representative for the MYRRHA reactor. • Infinite pin bundle: This reference quasi-direct numerical simulation profits from periodicity in all directions. It provides a detailed view into the flow field and in addition reveals details of the heat transfer from the rod bundle into the flow. Reference databases aim to serve the nuclear scientific community to validate engineering simulation approaches. The paper will introduce these reference databases, and how they have been used to validate RANS based turbulence modelling approaches within a mainly European context.

One of the most interesting technologies recognized as promising for the future of nuclear energy is the Accelerator Driven System (ADS). The purpose of the ADS is to function as an element of an integrated nuclear power enterprise comprising of conventional and advanced power reactors for energy production, and for reducing the radiotoxicity of the nuclear waste produced by these power reactors before entombment in a geologic repository ( Stanculescu, 2000 ). The use of new types of coolants such as molten lead or lead–bismuth eutectic alloy (LBE) represents another particular feature in the design concept of these reactors, since it permits one to take advantage of higher Heavy Liquid Metals (HLM) boiling temperature compared to water, leading to an improvement in thermal efficiency. Furthermore, it allows the reactor to be operated at a lower pressure, thus reducing the probability of a Loss-Of-Coolant Accident (LOCA) and, consequently, increasing reactor safety. The proposed use of relatively new types of coolants, especially in combination with new fuel types and cladding materials, demands specific attention to the thermal–hydraulics and core mechanics in normal and abnormal conditions. For that reason, a detailed analysis using system codes is necessary. In this paper, a new version of the ATHLET system code is tested, in which the physical properties of liquid metals like sodium, lead and LBE are implemented. The object of this study is the spallation loop of the MYRRHA facility, in which LBE is circulated by forced convection to remove the heat deposited by a proton beam. A detailed nodalization is set up for performing thermal–hydraulic calculations for both nominal conditions and accident scenarios in order to have a good characterization of the entire loop. The start-up transient has verified the correct removal of the heat generated in the target by the foreseen heat exchanger. During accidental transients, it was noted that the level of LBE in the beam line changes in agreement with preliminary studies on the target device. The pump failure test represents the most dangerous scenario, as the temperature in the target area will reach a very high value within a few seconds after the blockage of the pump impeller. The results obtained have subsequently been compared to the ones achieved with previous numerical simulations which were performed using a version of RELAP5/mod 3.3 system code that was modified at the University of Pisa to account for LBE properties. The comparison has confirmed the capability of the new version of the ATHLET code for the analysis of hydraulic circuits cooled by liquid metals, although both codes use different heat transfer correlations.

Thermal-hydraulics is recognized as a key scientific subject in the development of different innovative nuclear reactor systems. In Europe, thermal-hydraulic issues for innovative reactors are the subject of the Thermal-Hydraulics of Innovative Nuclear Systems project (THINS) which is sponsored by the European Commission and runs from 2010 until 2014. This paper will describe the status of the on-going activities in this project. The project addresses the main identified thermal hydraulics issues for innovative nuclear systems taking into account existing experimental data and producing new experimental data where needed. The overall objectives of the THINS project are the development and validation of new physical models, improvement and qualification of numerical engineering tools and their application to innovative nuclear systems. A specific part of the project is devoted to education and training in order to disseminate the state-of-the-art scientific knowledge gained within the project which will be attractive to students and (young) researchers in the field of thermal-hydraulics internationally.

Turbulence-induced vibration is typically considered as a type of vibration with one-way coupling between the fluid flow and the structural motion: the turbulence creates an incident force field on the structure, but the structural displacement does not influence the turbulence. It is however challenging to measure the turbulence forcing function experimentally. In this article, the forcing function in annular flow is computed by means of Large-eddy simulations. The pressure spectrum is applied to the inner cylinder and the resulting vibration is computed. It is shown that the commonly used multiplication hypothesis does not hold for the present results. The computed spectrum showed an upper limit to the coherence length. The results of these computations are compared to experimental results available in literature and to semi-empirical models. The predicted displacements compared well with experimental results.

The objective of this work is to assess the behavior of fuel redistribution in heavy liquid metal nuclear systems under fuel pin failure conditions. Two different modeling approaches are considered using Computational Fluid Dynamics (CFD) codes and a system code, applied to the MYRRHA facility primary coolant loop version 1.4. Two different CFD models are constructed: the first is a single-phase steady model prepared in ANSYS Fluent, while the second is a two-phase model based on the volume of fluid (VOF) method in STARCCM+ to capture the upper free-surface dynamics. Both use a Lagrangian tracking approach with oneway coupling to follow the particles throughout the reactor. The system code SIMMER-IV is used for the third model, without neutronic coupling. Although limited regarding the fluid dynamic aspects compared to the CFD codes, comparisons of particle distributions highlight strong similarities despite quantitative discrepancies in the size of fuel accumulations. These disparities should be taken into account while performing the safety analysis of nuclear systems and developing strategies for accident mitigation.

In this paper the modal characteristics of a flexible cylinder in turbulent axial flow are investigated with partitioned fluid-structure interaction simulations. In these simulations a computational fluid dynamics calculation to resolve the flow field is coupled with a computational structure mechanics calculation to compute the structural behavior. The cylinder is initially deformed according to an eigenmode in vacuo and then released. From this free vibration decay of the cylinder in the turbulent axial flow, modal characteristics are determined. To assess the accuracy of these calculations, the same configuration is computed as in an experiment with a solid brass cylinder mounted in a water-conveying pipe. The natural frequency appears to be relatively insensitive to an increase in flow velocity in this case. Both experiments and computations show the same trend of slightly decreasing natural frequency with increasing flow velocity. The damping, on the other hand, is very sensitive to the flow velocity. A change in flow velocity from 10m/s to 30 m/s results in a modal damping increase from 1.4% to 3.0%. Changes in molecular viscosity due to temperature differences had only a small effect on modal damping.

In this paper the vibration behavior of a flexible cylinder subjected to an axial flow is investigated numerically. Therefore a methodology is constructed, which relies entirely on fluid–structure interaction calculations. Consequently, no force coefficients are necessary for the numerical simulations. Two different cases are studied. The first case is a brass cylinder vibrating in an axial water flow. This calculation is compared to experiments in literature and the results agree well. The second case is a hollow steel tube, subjected to liquid lead–bismuth flow. Different flow boundary conditions are tested on this case. Each type of boundary conditions leads to a different confinement and results in different eigenfrequencies and modal damping ratios. Wherever appropriate, a comparison has been made with an existing theory. Generally, this linear theory and the simulations in this paper agree well on the frequency of a mode. With respect to damping, the agreement is highly dependent on the correlation used for the normal friction coefficients in the linear theory.

Excessive fuel assembly vibrations in nuclear reactor cores should be avoided in order not to compromise the lifetime of the assembly and in order to prevent the occurrence of safety hazards. This issue is particularly relevant to new reactor designs that use liquid metal coolants, such as, for example, a molten lead-bismuth eutectic. The flow of molten heavy metal around and through the fuel assembly may cause the latter to vibrate and hence suffer degradation as a result of, for example, fretting wear or mechanical fatigue. In this paper, we demonstrate the use of optical fiber sensors to measure the fuel assembly vibration in a lead-bismuth eutectic cooled installation which can be used as input to assess vibration-related safety hazards. We show that the vibration characteristics of the fuel pins in the fuel assembly can be experimentally determined with minimal intrusiveness and with high precision owing to the small dimensions and properties of the sensors. In particular, we were able to record local strain level differences of about 0.2 mu epsilon allowing us to reliably estimate the vibration amplitudes and modal parameters of the fuel assembly based on optical fiber sensor readings during different stages of the operation of the facility, including the onset of the coolant circulation and steady-state operation.

International audience; The windowless target electron beam experimental irradiation (WEBEXPIR) program was set-up as part of the MYRRHA/XT-ADS R&D effort on the spallation target design to investigate the interaction of a proton beam with a liquid lead–bismuth eutectic (LBE) free surface. In particular, possible free surface distortion or shockwave effects in nominal conditions and during sudden beam on/off transient situations, as well as possible enhanced evaporation were assessed. An experiment was conceived at the IBA TT-1000 Rhodotron, where a 7 MeV electron beam was used to simulate the high power deposition at the MYRRHA/XT-ADS LBE free surface. The geometry and the LBE flow characteristics in the WEBEXPIR set-up were made as representative as possible of the actual situation in the MYRRHA/XT-ADS spallation target. Irradiation experiments were carried out at beam currents of up to 10 mA, corresponding to 40 times the nominal beam current necessary to reproduce the MYRRHA/XT-ADS conditions. Preliminary analyses show that the WEBEXPIR free surface flow was not disturbed by the interaction with the electron beam and that vacuum conditions stayed well within the design specifications.

SCK•CEN, the Belgian Nuclear Research Centre, is designing an accelerator-driven system (ADS), MYRRHA, which aims to serve as a basis for the European experimental ADS and to provide protons and neutrons for various R&D applications. It consists of a proton accelerator that delivers a 350 MeV, 5 mA proton beam to a liquid lead-bismuth eutectic (LBE) spallation target, which in turn couples to an LBE-cooled, subcritical fast-spectrum core in a pool-type configuration. The liquid metal flow pattern in the lower part of the MYRRHA pool vessel needs to be investigated in order to assess the details of recirculation and stagnant zones for adequate coolant flow and sufficient physico-chemical mixing as well as to judge the scaling of flow down to a model that can be handled experimentally. To this end, three-dimensional (3-D) computational fluid dynamics calculations were performed by NRG in collaboration with the MYRRHA team.

The Belgian Nuclear Research Centre (SCK•CEN) is at the forefront of heavy liquid-metal (HLM) nuclear technology worldwide with the development of the Multi-purpose hYbrid Research Reactor for High...

External fluid flow has a number of effects on the dynamics of a submersed structure: e.g., a solitary cylinder exposed to an external flow experiences added mass and damping due to the presence of the surrounding fluid. At high axial flow velocities relative to the stiffness of the cylinder, coupled instabilities such as flutter and divergence occur. Compared to a solitary cylinder, a cluster of cylinders also experiences inter-cylinder coupling: pressure perturbations in the fluid due to the movement or acceleration of one cylinder force another cylinder to move. Consequently, the different cylinders can move in organized patterns. In this contribution, modal characteristics of a 7-rod bundle will be predicted by linear theory as well as by coupled CFD–CSM (Computational Fluid Dynamics — Computational Structure Mechanics) calculations. In the first part, fluid forces which lead to coupling of motion are computed with classical potential flow theory and URANS (Unsteady Reynolds-Averaged Navier–Stokes). Those forces are divided in a contribution in phase with the acceleration and a contribution in phase with the velocity of a cylinder. In the second part, modal characteristics of a 7 cylinder bundle are computed with coupled CFD–CSM simulations. The initial perturbations, which are required for the time-domain simulations come from a simplified structural model, with potential flow coupling between cylinders. The results are compared to linear theory. In the final part, approximations are proposed to predict upper and lower bounds of eigenfrequencies and damping, using calculations with only one cylinder.

This paper investigates the turbulence-induced vibration of a circular beam in annular pipe flow. Vibrations induced by turbulence are one of the causes of fatigue and fretting wear in process environments. Although the small-scale vibrations are normally not leading to immediate failure of structural components, they typically result in long term damage. To predict the amplitude of these subcritical vibrations, current methods require an accurate description of the incident pressure field. However, measurements of cross-spectral pressure fields in annular geometries are rare. Models to describe the pressure field have a tendency to provide only descriptive information, after a series of experiments have been performed. Therefore this paper aims to predict the pressure field numerically, by means of wall-resolved large-eddy simulations. In order to validate this approach the flow field of an experiment available in literature is computed. In the conditions simulated, water is flowing at 10 m/s in an annulus with a hydraulic diameter of 1.27cm. Pressure correlations obtained from the computations are compared to descriptive models such as the Corcos and Chase models. The numerical power spectra are also compared to experimental spectra.

The current paper describes the loss of flow (LOF) transient investigated in the MYRRHA reactor by the means of Computational Fluid Dynamics. This scenario is starting from the nominal operation case then the two pumps stop simultaneously. An unsteady solution with resolved interface was considered with calculating conjugate heat transfer through the relevant structures. Due to a postulated event (e.g. loss of the electric grid) the pumps are not powered anymore stops. After the detection of the problem (temperature difference above the core rises with 20 degree) the reactor power is stopped by the safety rods (delay of 1 s). The fuel elements, however, continue to generate residual heat according to the decay heat curve. Due to the loss of the pumps, the pressure difference between the cold and the hot plenum is decreasing, which result in a gravitational flow equilibrating the two free surfaces to the same level. The objective of the work was to determine the flow through the core during the coast down of the pumps and eventual flow reversal into the pump/heat-exchanger box due to the gravitational flow. The simulation revealed that after losing power, the LBE flow reverses into the pumps in less than 0.1 s according to the simulations. In the core there is a brief moment of reverse flow, too, but only after the core is scrammed, therefore, the loss of cold LBE flow is not causing overheat. Once the core is scrammed, the position of the maximum temperature in the system shifts to the Above Core Structure, where the residual hot plume rising from the core impinges to the Above Core Upper Closure. The levels of the lower and upper plenum equilibrate roughly 20 s after the pump failure event.

In recent years, extensive thermal-hydraulic experimental tests have been performed on the lead-bismuth eutectic (LBE)–cooled, wire-wrapped fuel assembly (FA) of MYRRHA. These thermal-hydraulic tests were performed using FA mock-ups in large-scale LBE experimental test facilities at SCK•CEN (Belgium), ENEA (Italy), and Karlsruhe Institute of Technology (Germany). The FA pressure drop characteristics and flow-induced vibration (FIV) characteristics were tested with a full-scale 127-pin mock-up test section. The existing pressure drop correlations of Rehme and of Cheng and Todreas (simplified model) predict the experimental pressure drop data very well and are considered suitable for use in the design and safety analysis of the MYRRHA system. FIVs are very limited in the wire-wrapped bundle, and fuel pin fatigue damage from vibration during operation is not expected. Further analysis and testing are required to determine if damage from fretting corrosion could be expected. Heat transfer characteristics of the FA were investigated experimentally in two separate 19-pin heated rod test sections cooled by LBE. The existing Kazimi-Carelli correlation predicts the global average Nusselt numbers very well, but the correlation is not developed to capture local hot spots. For the FA safety analysis, to further determine operational safety margins, a hot-spot factor is defined and analyzed to determine the hot-spot temperature penalty.